Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu.D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

doi:10.1134/s0040601514080084

Cited by 7 publications

(2 citation statements)

References 7 publications

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“…Let us address the list of problems required to be investigated for substantiating the VVER SCP (supercritical pressure) concept and for initiating the design of pilot reactor with 30-MW power, which seems to be sufficiently standard and obvious in designing any type of nuclear reactors (Ryzhov et al 2009, Kalyakin et al 2014, Alekseev et al 2015, Schulenberg et al 2014 Quite probably these will be chrome-nickel alloys (17Cr, 13Ni, 2Mo) for which it is necessary to implement estimation of their swelling for respective intensities of irradiation (dpa).…”

Section: Scientific and Technical Problems Requiring Resolution In Thmentioning

confidence: 99%

Generation IV supercritical water-cooled nuclear reactors: Realistic prospects and research program

Kirillov¹,

Bogoslovskaya²

2019

NUCET

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Existing conditions make possible obtaining information that being discussed openly by wide scientific community could help outlining or even establishing the expediency of a particular area of present and future research. Use link http://www.sciencedirect.com to learn about the topics or areas that most attract researchers from different countries. The Generation IV International Forum (GIF-IV) established in January 2000 has set a goal to improve the new generation of nuclear technologies in the following areas: stability, safety and reliability, economic competitiveness, proliferation resistance and physical protection. The purpose of the present publication is to prepare a discussion of one of the directions of development of fourth-generation NPPs, the groundwork for which has already been laid in thermal power engineering in various countries. The number of papers published annually on this topic is the largest among other similar topics dedicated to nuclear power plants of the fourth generation. Judging from the operating experience of existing nuclear power plants using water as a coolant, it can be ascertained that the tendency of building water-cooled nuclear power plants will remain during the next 30 to 50 years. During the present stage the task in the development of alternative types of reactors will be limited to demonstration of their performance and acceptability for future power engineering and the society. The project of supercritical water-cooled reactor is based on the operating experience of VVER, PWR, BWR reactors (more than 14,000 reactor-years); many years of experience accumulated in operating fossil thermal power plants (more than 400 power units; 20,000 years of operation of power units) using supercritical (25 MPa, 540°C) and super-supercritical (35–37 MPa, 620–700°C) water steam. In Russia more than 140 supercritical pressure units are currently in operation. Numerical calculation and design of supercritical water-cooled reactor (similarly to BR-10 reactor) will allow not only training personnel for future development of this technology, but will also help revealing the most difficult points requiring experimental confirmation with application of independent test facilities, as well as formulating the plan of first priority experimental studies. Knowledge accumulated over the last 10 years in the world allows the following: further specifying the already developed concept; developing a plan of specific priority studies; compiling task order for designing small-power pilot VVER SKP-30 reactor (30 MW-th). The scope of problems that are to be solved to substantiate VVER-SCP reactor and commence designing an experimental reactor with thermal capacity of 30 MW is the same as that in developing any type of nuclear reactor: physics of the reactor core; material related matters (primarily concerned with the reactor pressure vessel, fuel, and fuel rod cladding); thermal hydraulics of rod bundles in the near- and supercritical areas; water chemistry at supercritical pressure; corrosion of materials, development of safety systems. Research must be carried out both in static conditions and under irradiation. The absence in Russia during the extended time period of approved program with allocation of appropriate funding and preservation of the existing status during the coming two or three years will lead to the situation when Russia will be hopelessly lagging behind in the development of SCWR technology.

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Section: Scientific and Technical Problems Requiring Resolution In Thmentioning

confidence: 99%

Generation IV supercritical water-cooled nuclear reactors: Realistic prospects and research program

Kirillov¹,

Bogoslovskaya²

2019

NUCET

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“…CFD -(Computational Fluid Dynamics) розрахункова гідродинаміка; SCWR -(Super Critical Water-cooled Reactor) надкритичний реактор з водяним охолодженням; SST -(Shear Stress Transport turbulent model) модель турбулентності переносу зсувних напруг. енергетиці потребує глибокого знання особливостей теплогідравліки в реакторах з водою надкритичних параметрів (SCWR) [1,2].…”

Section: скороченняunclassified

CFD Modeling of Heat Transfer Under Flow of Supercritical Parameters Water in Vertical Bare Tubes

Fialko

Pioro

Prokopov

et al. 2018

ihe

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The results of the CFD modeling of heat transfer under condition of the upstream flow of supercritical water in vertical bare tubes are presented with a change in the wide range of mass flux and heat flux at the channel wall. The results of comparison of experimental and calculated data for different heat transfer regimes - improved, normal and deteriorated - are presented. The results of the CFD modeling relating to the spatial distribution of physical properties of supercritical water are analyzed.

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Heat-Exchange Regularities and Thermohydraulic Code Verification for Supercritical Pressure Water Cooled Reactors

Kirillov

2021

At Energy

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Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

Cited by 7 publications

References 7 publications

Generation IV supercritical water-cooled nuclear reactors: Realistic prospects and research program

Generation IV supercritical water-cooled nuclear reactors: Realistic prospects and research program

CFD Modeling of Heat Transfer Under Flow of Supercritical Parameters Water in Vertical Bare Tubes

Heat-Exchange Regularities and Thermohydraulic Code Verification for Supercritical Pressure Water Cooled Reactors

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